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3 High-Level Radioactive Waste The majority of HLW is the highly radioactive waste stream from reprocessing of SNF (see Sidebar 3.1 for a discussion of reprocessing methods). Other highly radioactive material, such as reactor compartments from nuclear-powered submarines, also fit in this category in Russia, although the United States considers these to be a different class of waste (See Sidebar 1.1). 3.1 HIGH-LEVEL RADIOACTIVE WASTE IN THE RUSSIAN FEDERATION The total activity of radioactive wastes accumulated at Minatom enterprises, not accounting for wastes injected deep underground, exceeds 2×109 Ci (see Tables 3.1 and 3.2). 3.1.1 Production Association “Mayak” Nowadays, the reactor division of PA “Mayak” operates two 1,000 MWth reactors, Ruslan and Lyudmila, producing radionuclides both for military and civilian purposes. Five uranium graphite reactors were shut down between 1987 and 1991. Production of the weapons-grade plutonium at PA “Mayak” ceased in 1987. The radiochemical plant operation started in 1976 and since then its staff has reprocessed spent fuel from different power reactors, as well as from transport and research reactors. During operation of the RT-1 plant, 2,380 tons of spent fuel have been received from domestic and foreign power plants for reprocessing. Prior to RT-1, PA “Mayak” operated the first radiochemical plant, known as Plant “B,” to process irradiated targets from the first production reactor, reactor “A.” Plant “B” operated from December 1948 until the 1960s. In 1959, Plant “BB” was brought on
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SIDEBAR 3.1: REPROCESSING The RT-1 plant extracts plutonium and uranium from SNF using the PUREX process. PUREX is an aqueous process in which the declad and crushed fuel matrix is dissolved in nitric acid yielding a feed to a multi-stage cascade that extracts and strips uranium and plutonium (U and Pu). The solutions contain high concentrations of particulates (graphite, silicon, and others) in suspension, so the solution is clarified with filters and organic flocculants. U and Pu (and, currently, neptunium) are extracted with an organic solvent, tributyl phosphate (TBP) in a saturated hydrocarbon similar to kerosene, which leaves behind essentially all of the other constituents. Nitric acid with a reducing agent strips the Pu from the TBP and contact with dilute nitric acid strips the U back into the aqueous phase. RT-1 repeats this process twice. The PUREX process is very effective at recovering nearly all the U and Pu, leaving insignificant levels (one part in one hundred million) of residual contamination with fission products and minor actinides, but the process generates large amounts of waste and cannot separately recover (fractionate) other constituents for recycling or specialized disposal. Radiolysis and chemical processes degrade the TBP, which must be continuously purified, and this purification process also generates large volumes of waste. Equipment choices, such as centrifugal contactors, can achieve some reductions in volume by promoting faster reaction resulting in less exposure and fewer radiolytic effects, but the clarification and extraction processes still generate large amounts of waste. As processing of SNF continues in Russia, and particularly if the program is to expand to accept VVER-1000 and RBMK SNF and SNF of international origin, the Russian Federation must examine and pursue ways to improve the radiochemical processes employed to carry this out. Other schemes might improve the characteristics of the process with respect to the environment, proliferation (theft of special nuclear material), safety, and economics. These aspects all must be examined. Several enhancements and alternatives to PUREX are close to being ready for production-scale deployment. These include UREX, TRUEX, volatilization using AIROX; dry reprocessing technologies using flourination, or electrochemical separation in molten salt; and several others. UREX is a modification of the front end of the PUREX process that uses the reagent AHA (acetohyroxamic acid) to complex Pu and reduce its valence so that the Pu will remain in the aqueous phase when the uranium is extracted into TBP, as in PUREX. This allows high-purity recovery of the U from SNF, leaving the Pu with the minor actinides and fission products. UREX is attractive in systems that keep Pu and the minor actinides together for proliferation resistance and actinide burning (see, e.g., the integral fast reactor concept with pyroprocessing, or transmutation in general with other partitioning techniques).
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The TRUEX process uses a strong chelating agent to extract all the actinides except uranium, neptunium, and plutonium from an immiscible organic solvent, such as TBP. TRUEX can recover americium, curium, and higher actinides from the PUREX HLW stream, although it also extracts several lanthanides, iron, and zirconium at the same time (NRC 1996a). AIROX is a dry process that removes fission products from SNF by volatilizing them during oxidation and reduction cycles, taking advantage of the fact that when oxidized in O2, UO2 SNF forms a less dense matrix of U3O8. The volume increase cracks the fuel and heat drives the volatile and semi-volatile fission products from the fuel. U3O8 is then reduced to UO2 by H2 and the cycle is repeated. Pu is never separated from the U and minor actinides in the SNF, and can be loaded with more fissile material and refabricated into fuel. Flouride volatility processing is based on the fact that U, Pu, and Np form volatile flourides, but few other elements do. The Midwest Fuel Recovery Plant in Morris, Illinois, in the United States (see Section 1.2) was designed to use both flouride volatility and solvent extraction methods, but the plant never operated because of faulty designs that made the plant infeasible to operate. Molten salt processes rely on the different thermodynamic and electrochemical properties of different elements when dissolved in ionic molten inorganic salts. The dissolved constituents of the SNF can be separated by volatility or by ionic transport, which can be driven by thermodynamic activity differences in different media or by potentials between electrodes. The chief advantage of molten salt processes is their resistance to radiation damage effects and consequently their effectiveness in processing intensely radioactive fuel that has been discharged from a reactor. (Oxide fuels typically must be reduced to metal before separations in these processes.) This technique has been recently tested in Russia in a semiproduction-scale at a research nuclear reactor, with the regenerated fuel reused for fabrication of fuel elements for fast breeder reactor BOR-60. Other processes based on solvent extraction (dicarbollide, crown ethers, supercritical fluids, and others), ion exchange and adsorption, membranes, precipitation, and others could improve steps in processing of SNF. (See Appendix D of NRC [1996a] for more details.) line to process production-reactor targets. A partially completed second line of Plant “BB” was found to be unnecessary for production of weapons materials, and the facilities were adapted to create the radioisotope plant, which processes targets from the isotope-production reactors (Cochran et al. 1995).
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TABLE 3.1 Generalized Data on the Amounts of Liquid Radioactive Waste (LRW) Currently Generated, Reprocessed, and Stored at the Minatom RF Enterprises Name of Enterprise Total Including: High-levelb Intermediate-level Ci 103 m3 Ci 103 m3 Ci 103 m3 LRW generated in 2001 Minatom RFa 1.06×108 3800 6.95×107 10.9 3.57×107 463 PA “Mayak” 7.05×107 1170 6.95×107 10.8 1.01×106 21.1 LRW reprocessed during 2001 Minatom RF 4.92×107 1630 4.68×107 17.1 2.32×106 17.5 PA “Mayak” 4.68×107 438 4.68×107 17.1 7.08×104 2.26 LRW accumulated by the end of 2001 Minatom RF 1.84×109 469000 3.51×108 30.2 1.49×109 11400 PA “Mayak” 4.76×108 412000 3.51×108 29.2 1.22×108 473 a Minatom RF includes PA “Mayak,” Krasnoyarsk MCC, and the SCC. b No official data are available on liquid high-level waste from Krasnoyarsk MCC and the SCC. NOTE: Does not include wastes disposed deep underground. SOURCES: Shatalov (2002), Minatom (2002). Some numbers have been rounded.
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TABLE 3.2 Generalized Data on the Amounts of Solid Radioactive Waste (SRW) Currently Generated, Reprocessed, and Stored at the Minatom RF Enterprises Name of Enterprise Total Including: Spent sealed radioactive sources High-level Intermediate-level Ci 103 tons Ci 103 tons Ci 103 tons Ci Number SRW generated in 2001 Minatom RF 1.28×106 863 1.28×106 0.905 2.76×103 6.06 1.91×105 14100 PA “Mayak” 1.05×106 1.75 1.05×106 0.529 1.69×103 0.072 6.54×104 1190 SRW treated in 2001 Minatom RF 5.16 2.68 N/A N/A 2.84 1.85 N/A N/A SRW accumulated by end of 2001 Minatom RF 2.29×108 57100 2.29×108 126 3.03×104 815 5.03×106 68000 PA “Mayak” 2.24×108 309 2.24×108 43.8 3.27×103 100 4.78×106 17900 NOTE: Does not include wastes disposed deep underground. SOURCES: Shatalov (2002), Minatom (2002). Some numbers have been rounded.
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For radiation protection of workers, incoming SNF is kept for five to seven years in cooling pools during decay of the short-lived isotopes. The SNF is then chopped up and dissolved in concentrated nitric acid, from which elements are extracted and separated using organic solvents. The final stage of the process produces highly purified metal oxides and their salts. All processes are executed with remotely controlled equipment. The residual solution is subjected to further treatment for extraction of commercially used isotopes. In the Russian Federation, the uranium extracted from power-reactor SNF is then mixed with more highly enriched uranium from propulsion-reactor and research-reactor SNF and then sent, in the form of uranyl nitrate, for fabrication of RBMK fuel. Plutonium dioxide is transferred to the storage facility located within the plant territory. Plutonium is supposed to be used in the future for fabrication of fuel for fast-neutron reactors. As a result, a closed nuclear fuel cycle could utilize all of the uranium and plutonium. HLW in the Environment The largest outflow of radioactive waste into the environment was during the early operational period of the first radiochemical plant, known as Plant “B” (February 1949). In accordance with the technology adopted at that time, the waste water from Plant “B” was poured directly into the Techa River, at an outflow up to 1,000 Ci per day. The total amount of radioactivity released into the Techa River between 1949 and 1956 was 2.75× 106 Ci (Mokrov 1996). The river flood plain was polluted over a distance of 100 km. From 1958 on, the largest amount (more than 1.2×108 Ci) of liquid intermediate-level nuclear waste accumulated in Lake Karachai (Reservoir 9). In 1957, a chemical explosion in a liquid high-level waste tank resulted in the release of 2×106 Ci over a large area (and about 1.8×107 Ci deposited in the immediate vicinity of the tank), which led to the formation of the relatively narrow but long East Urals Radioactive Trail (EURT) with a pollution density of 2 Ci Sr-90 per square kilometer over 1,000 km2, mostly overlying the EURT (Joint Norwegian-Russian Expert Group for Investigation of Radioactive Contamination in the Northern Areas 1997).
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In 1967 wind dispersed 600 Ci of radioactive compounds from Karachai Lake and its banks, resulting in contamination of a 30 km2 territory at a density of 2 Ci Sr-90/km2. In order to prevent further aerosol dispersion from the lake surface and shores, the area of the lake was diminished from 51 ha in 1962 to 15 ha in 1993 by filling it with gravel and hollow concrete blocks. Liquid radioactive waste dumped into Karachai Lake caused contamination of underground water. The total amount of radioactive solutions supplied by the lake to groundwater is about 3.5×106 m3, which includes ~7×104 Ci Sr-90; 2×104 Ci-137 Cs; 6.6×105 Ci Ru-106; 1×105 Ci 3H; and a considerable amount of uranium, neptunium, and plutonium. The dynamics of radionuclide dispersion in groundwater is an urgent scientific problem (Drozko et al. 1996). During the 45-year period of nuclear weapons production, PA “Mayak” accumulated 6×108 Ci of liquid HLW. This waste could not be vitrified because of its complex chemical composition. About 2.2×108 Ci of solid HLW was stored in 24 reinforced concrete surface structures and about 3×103 Ci of intermediate-level waste and low-level waste in 200 near-surface land-fills. More than 3×108 m3 of contaminated water was accumulated in industrial ponds created in the Techa River valley (mainly ponds 10 and 11 with areas of 19 km2 and 44 km2, volumes of 7.6×107 m3 and 2.3×108 m3, and 1.1×105 Ci and 3.9×104 Ci, respectively). The PA “Mayak” area currently contains ~8×108 Ci of radioactive waste in various forms, which is clearly a serious environmental hazard, primarily because of the possible outflow of radionuclides into the Techa-lset’-Tobol-lrtysh-Ob’ stream system that drains into the Kara Sea. 3.1.2 Krasnoyarsk Mining and Chemical Combine Krasnoyarsk Mining and Chemical Combine (MCC) is located underground at the closed administrative area Zheleznogorsk, and occupies 360 km2 on the right bank of the Yenisey river.1 1 Information on wastes discharged to surface waters at the Krasnoyarsk MCC is available in a paper by Georgievsky (2001).
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The enterprise includes a complex of production and supporting facilities the core of which are the reactor production facility and radiochemical plant (Stepanova 1996). There is also a storage facility for heat-generating irradiated-fuel assemblies and a factory for nuclear waste treatment. The reactor facility now includes one closed-circuit uranium graphite reactor. This reactor powers an underground station that provides heating and hot water to Krasnoyarsk-26. The radiochemical plant reprocesses irradiated fuel from the reactor. The MCC currently stores 6,500 m3 of high-level, sludge-like, legacy nuclear waste from its weapons-grade plutonium production program. The total activity of this waste is 1.3×108 Ci. Ninety percent of the waste is in the tank complex of the Radiochemical Plant. Located in a rock massif, the underground storage complex comprises nine stainless-steel-lined tanks, each with a capacity of 3,200 m3. The remaining ten percent of the waste is in a subsurface storage complex consisting of eight toroidal-shaped storage tanks each with a capacity of 8,500 m3. Four of these subsurface tanks are stainless steel lined and four are lined by carbon steel with epoxy coating. High-level liquid wastes are initially somewhat homogeneous. During their storage, however, small amounts of silicic acid precipitates are formed that readily absorb radionuclides and that settle out as sludge in the tanks. A multiyear study has shown that the sludge solids basically consist of metal hydroxides (steel corrosion products and aluminum); polymerized forms of silicic acid; oxides of niobium (V) and manganese; ferrocyanides of nickel and cesium; and residues of ion-exchange resins. The sludge solids also contain significant amounts of uranium and plutonium. The main radioactive constituents of the wastes are isotopes of U, Pu, Np, Th, Zr, Nb, Ce, Cs, and Sr. Intermediate- and low-level liquid wastes are directed to deep-injection disposal into hydraulically isolated, permeable horizons at the injection site “Severny” located 12 km to the Northeast from the main production zone of the enterprise (Compton et al. 2000; Parker et al. 1999; Parker et al. 2000; Malkovsky et al. 1999). The MCC has disposed of waste at “Severny” since 1967. As of 1995, about 5 million m3 of LRW with total activity of about 260×106 Ci (decay corrected to 1995) had been injected into two deep aquifers.
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3.1.3 Tomsk Siberian Chemical Combine The Siberian Chemical Combine (SCC) is located near the town Seversk (Tomsk-7). The SCC was commissioned in 1953. The SCC area is 192 km2 within an observation zone of 1,560 km2. Within the observation zone area are the town of Seversk, several settlements, and a part of the city of Tomsk. At present, SCC is a complex for production of plutonium, uranium, and transuranium elements. It has several production facilities among which are a reactor plant running graphite-moderated reactors (ADE-4,5) designed for production of weapons plutonium and electric power; a gas-centrifuge uranium enrichment plant; a sublimate plant for production of uranium oxide and uranium hexafluoride; a radiochemical plant for reprocessing of irradiated standard blocks2 for production of uranium and plutonium salts; and chemical-metallurgical facilities for fabrication of nuclear materials. SCC also has facilities for storage of radioactive materials, including materials from nuclear warheads that were recently placed into the specialized buildings (Security Council 1994). Production of plutonium, uranium, and transuranium elements at the SCC results in generation of considerable amounts of liquid, solid, and gas-aerosol radioactive wastes. SCC has 50 storage facilities of liquid and solid radioactive wastes in its territory, including sites for deep injection of liquid radioactive wastes, which up to now have not impacted the biosphere, but present potential hazards for the environment. The total amount of liquid radwaste disposed of in the deep geological formations is assessed to be 4×108 Ci, and the amount in surface storage facilities is approximately 1.25×108 Ci. 3.1.4 Dimitrovgrad Scientific Research Institute of Nuclear Reactors The Scientific Research Institute of Nuclear Reactors (NIIAR) is located 5 km to the west of the town of Dimitrovgrad, Ul’yanovskaya oblast. NIIAR is a large research center with seven operating research and power reactors, where studies are performed on reactor materials science, fuel elements and assemblies, techniques for dry SNF reprocessing, fabrication of pluto- 2 A “standard block” is a fuel element in aluminum cladding manufactured from natural uranium metal.
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nium-uranium mixed-oxide fuels, techniques for producing transuranium elements, and fabrication of ionization sources. The pilot-industrial site (PIS) for disposal of the non-technological wastes from the research installations is located on the territory of an industrial zone near the purification facilities where pretreatment of liquid wastes is carried out before their deep injection disposal (Rybal’chenko et al. 1994). The PIS has operated since 1966. The horizons selected to host injected wastes are at depths between 1,440 and 1,550 meters (Horizon III) and between 1,130 and 1,410 meters (Horizon IV), with effective thicknesses of 35 and 80 meters, respectively. Injection horizons are hydraulically isolated above and below by low-permeability layers. The groundwater velocity in the geologic formation is rather difficult to assess, but estimates do not exceed one centimeter per year. 3.1.5 Obninsk Institute of Physics and Power Engineering The Institute of Physics and Power Engineering (IPPE) is located in the town of Obninsk, Moscow oblast, on the left bank of the Protva River about 100 km southwest of Moscow. In the town, there are several institutions with potential radiation hazard to the environment, among which the IPPE and a branch of the Physico-chemical Institute are the main ones. The main type of impact exerted on the environment by the local institutions are radionuclide gas-aerosol emissions to the atmosphere and radionuclide discharges with the waste waters to the Protva River, as well as radionuclide contamination of the subsurface groundwaters. Since IPPE began operations, 1,100 m3 of liquid radioactive wastes with a total activity of 1.63×105 Ci, and 2.3×104 m3 of solid radioactive wastes with a total activity 0.14×105 Ci have accumulated. The radionuclide inventory includes Cs-137, Cs-134, Mn-54, Co-60, U-235, Pu-239, and others. 3.1.6 Production Association “Sevmashpredpriyatie” The Production Association “Sevmashpredpriyatie” is located in the town of Severodvinsk, Arkhangelsk oblast, on the shore of the Dvinsk Gulf of the White Sea, 35 km west of the town of Arkhangelsk. The “Sevmashpredpriyatie” production association
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carries out construction and repair of nuclear-powered ships. Also located there are the dockyard, “Zvezdochka” with a facility for interim storage of radioactive waste, and a nearby base, where testing and partial dismantling, salvage, and disposition of nuclear-powered submarines are carried out.3 3.2 HIGH-LEVEL RADIOACTIVE WASTE IN THE UNITED STATES “The highly radioactive waste material resulting from the reprocessing of spent nuclear fuel (SNF), including liquid waste produced directly in reprocessing and any solid material derived from such liquid waste that contains fission products in sufficient concentrations; and other highly radioactive material that is determined, consistent with existing law, to require permanent isolation” (DOE 2001a). 3.2.1 Defense High-Level Radioactive Waste Nearly all (over 97 percent by volume) of the HLW in the United States was generated by chemical processing of irradiated targets or fuel from production reactors at two sites (the Hanford Site and SRS) as part of the nuclear weapons material production programs. Relatively small amounts (by volume) were produced in reprocessing of SNF from naval reactors at the INEEL, and in reprocessing of commercial SNF at the Western New York Nuclear Service Center (now called the West Valley Demonstration Project).4 Much smaller quantities are still being generated in processing of “at risk” SNF at the SRS and Argonne National Laboratory-West. Table 3.3 summarizes the volumes of HLW in tanks and the numbers of canisters of vitrified HLW stored at the sites, as of 1999. The plants that generated the majority of the HLW used the PUREX process to extract plutonium and uranium from the SNF.5 In these processes, the fuel is typically chopped up, the fuel clad- 3 These sites are listed here to identify some of the other sites involved in handling and storage of SNF and HLW in Russia. 4 The Savannah River Site also processed a very small amount of commercial SNF. 5 The Hanford Site used other processes before it built the PUREX Plant. Argonne National Laboratory-West has a research-scale electrochemical processing system, but it has processed only a small quantity of fuel.
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TABLE 3.3 Quantities of HLW Stored at Sites in the United States Site HLW in Tanks (cubic meters) Vitrified HLW (canisters) Total Radioactivity (×106 Ci) Percent of Total Volume Percent of Total Radioactivity Hanford Site 200,000 0 384 58.9 15.8 Savannah River Site 130,000 719* 1,730 38.3 71.0 Idaho National Engineering & Environmental Laboratory 9,360 0 300 2.8 12.3 West Valley Demonstration Project 109** 241** 23.3 <0.1 1.0 Total 339,000 960 2,430 100 100 * Current number is 1,337 as of October 2002. ** HLW from the tanks at West Valley has been vitrified in 275 canisters. Residual HLW encrusted on the tanks is being characterized and sluiced. SOURCE: DOE (2001b).
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other soluble radionuclides7 were discharged directly into the subsurface.8 These constitute the largest discharges to the ground at Hanford, by radioactivity (approximately 5 million curies, in the second row of Table 3.4). Groundwater under more than 100 square miles (260 square kilometers) of the Hanford Site is contaminated above drinking-water standards with radionuclides and chemicals, including tritium, strontium-90, technetium-99, iodine-129, uranium, carbon tetrachloride, and chromium. Uranium and toxic chemicals also were discharged to the ground through drains in conjunction with plutonium recovery processes, and at least 67 of the 177 underground tanks at Hanford are known or suspected to have leaked HLW directly into the subsurface. Many of these tanks have exceeded their design lives. The total radionuclide input to the subsurface at Hanford from HLW operations is unknown but is probably on the order of a few million curies (NRC 2001b) (see Table 3.4). Some of the contaminants released or pumped into the ground have formed large underground plumes that are intersecting the Columbia River, but there were also direct discharges to the river. The largest sources of direct releases to the river were the eight “single-pass” production reactors. These reactors used treated river water as coolant and the neutron-activated constituents, carrying small amounts of fission products, were discharged back into the river. Heeb and Bates (1994) estimate that about 110 million curies were discharged to the river, although most of this was short-lived (half-lives on the order of days or less) and would not be considered HLW. Although SRS and INEEL have zones of contamination from a variety of sources, including buried wastes, little of this contamination is the direct result of leaks or releases of HLW. There has been some HLW leakage into the subsurface at the Idaho site from valves and a severed waste transfer line; the amount of leakage is on the order of a few tens of thousands of curies. At Savannah River a small amount of HLW (on the order of tens of liters) is known to have leaked into the subsurface from one tank, and several tanks have had leaks from their primary shell into the annular region between the shells. 7 Cesium was in the liquid waste stream during operation of the cascade of tanks until the 1950s. After that, cesium was precipitated in some tanks with ferrocyanide and the remaining supernate was still dumped in the ground. 8 HLW volumes at Hanford were reduced by about a factor of 10 by this chemical treatment/discharge process and evaporation.
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TABLE 3.4 Inventory of High-Level Waste in the 200 Area at the Hanford Site Waste Volumes (millions of gallons) Curies to Ground (millions)a Curies in Facilities (millions)a References HLW Generated 530 — — Agnew (1997) Direct Discharges to Soilb 120–130 0.065–4.7c — Waite (1991); Agnew (1997) Tank Leaks to Soild 0.75–1.5 0.45–1.8 — Waite (1991); ERDA (1975); Agnew (1997) Evaporator Condensates Discharged to Soil 280 0.003 — Agnew (1997); Hanlon (2000); Wodrich (1991) Cooling and Processing Water 400,000 Negligible — DOE (1992b,c) Cs and Sr Capsules — — 140 DOE (1996b) Appendix A, Table A.2.2.1 Tank Waste 54 210–220 Waite (1991); Agnew (1997); Hanlon (2000) Facilities — — 10e Gephart (1999) Totals 0.22–6.5 360–370 NOTE: Numbers are rounded to two significant digits from the values given in the references. The numerical ranges represent differences in estimating procedures and do not necessarily represent uncertainty ranges of the estimates themselves, which have not been determined, in part because the quality of the estimates are unknown. aQuantities are decay corrected to the mid-to-late 1990s. bAfter cascading through multiple tanks or after chemical treatment to remove cesium. cThe lower estimate is for Cs-137 and minor amounts of Sr-90 only. dEstimate does not include leaks from transfer lines and valves. eRadionuclides estimated to remain in plutonium production reactors and chemical separations facilities. SOURCE: NRC (2001b).
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3.2.3 High-Level Radioactive Waste from Processing Commercial Spent Fuel Two commercial reprocessing plants were built in the United States: the Nuclear Fuel Services facility near West Valley, New York, with a 300 MTHM per year design capacity, and the Midwest Fuel Recovery Plant in Morris, Illinois, also with a 300 MTHM per year design capacity. A third facility, the Allied General Nuclear Services plant in Barnwell, South Carolina, designed to process up to 1500 MT per year, was never completed. The Midwest Fuel Recovery Plant was completed but was found to profound design flaws in 1974, and was not put into operation, but is used as a storage site for SNF. The Nuclear Fuel Services plant (also called the Western New York Nuclear Service Center and later the West Valley Demonstration Project) primarily used the PUREX process, but also used the THOREX process for some thorium-bearing fuels, and began operating in 1966. The plant processed approximately 640 MTHM, roughly 60 percent of which was from the N-Reactor at Hanford, and the remainder was from commercial nuclear power plants. The facility shut down in 1972 to make modifications intended to seismically stabilize the facility and increase its capacity, but it never restarted. In addition to the fuel reprocessed at West Valley, a very small amount of commercial SNF, 0.7 MTHM, was reprocessed at SRS (EIA 1999a). During operations, the West Valley facility generated approximately 2,000 m3 of liquid high-level waste that was stored in two underground tanks: a 51- m3 stainless steel tank, and a 2,800-m3 tank made of carbon steel. Another identical set of tanks was left empty during operations, but has been used in treating the tank wastes. The program to vitrify liquid HLW at West Valley was completed in August 2002 with the production of the last of 275 vitrified logs. 3.3 END POINTS FOR HIGH-LEVEL RADIOACTIVE WASTE THAT IS NOT SPENT NUCLEAR FUEL9 At the PA “Mayak” plant, the liquid HLW is vitrified in the EP-500 electric ceramic melter. The phosphate glass fabricated in this melter is poured into containers and then moved to the temporary storage facility in the RT-1 plant vitrification area. The first 9 End points for SNF are addressed in Section 2.4.
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EP-500 melter lasted one year (there were problems with the current supply and coolant). The second melter was in continuous operation for six years (1995–2001). Over the course of this operation as much as 12,000 m3 of the high-level liquid waste were reprocessed to produce more than 2,300 tons of glass (total radioactivity of approximately 3×108 Ci), which are currently stored. The third EP-500 melter was put into operation in 2002. To end the practice of dumping liquid intermediate-level waste into open waters at PA “Mayak” (Lake Karachai and the Techa Ponds Cascade), PA “Mayak” is developing a technology for joint vitrification of high-level and intermediate-level waste in the EP-500 melters. Work has been done to develop an induction melter, with a “cold” crucible provided with an inlet direct-feed evaporator, to allow reprocessing of liquid waste of a wide range of compositions to produce materials with desirable properties (glass and mineral-like crystal matrices). By using different mineral-like matrices, stabilized compounds can be produced for a variety of waste forms. PA “Mayak,” however, currently has no plans to switch to a cold crucible melter and has a fourth melter of the current design already installed for use when the new third melter reaches its end of life. Thus, the newer technology is not yet in use. Metal radioactive waste (parts of irradiated fuel assemblies, fuel cladding, etc.) are sent for storage in specialized storage facilities. A technology for induction-slag remelting of such waste in the “cold” crucible has been developed to reduce the volume of metal radioactive waste by 5–6 times and to decontaminate the metal, thereby decreasing the residual activity by two or three orders of magnitude, raising the possibility of reuse. Fiberglass and gauze filters and adsorbers are used in the gas cleaning systems of the plant and the nuclear power plant to remove aerosols and iodine from gaseous releases. 3.3.1 Nuclear Waste Underground Disposal and Disposition in the Russian Federation Geological disposal of solid and solidified HLW is considered in Russia as being economically, technically, and ecologically the most attractive approach to completion of the nuclear fuel cycle. In accordance with previous decisions by Minatom, work on selection of the sites for HLW disposal and construction of SNF
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storage facilities are assigned and planned at the radiochemical enterprises: PA “Mayak,” Krasnoyarsk MCC, as well as in the regions of the nuclear Navy bases in the Russian Far East and Northwest. The first stage of developing repositories is construction of underground research laboratories at sites at the PA “Mayak,” at the Krasnoyarsk MCC, and in the Northwest region. Production Association “Mayak” At present, more than 2,000 metric tons of radioactive aluminophosphate glass from vitrification of liquid HLW with total activity about 3×108 Ci is stored at PA “Mayak.” In addition, liquid HLW with a total activity of about 3.77×108 Ci, which also is destined for vitrification, is currently stored in the special-purpose reservoirs. All vitrified HLW at PA “Mayak” is destined for underground disposal within the area of the enterprise’s sanitary-protection zone (SPZ). The SPZ territory is formed by volcanic rocks of andesitebasalt composition, which have effective physical and geochemical isolation capabilities.10 The rock massif is, however, cross-cut by numerous irregularly distributed faults of different scales. Within the fault zones, rocks are strongly technically disturbed and are characterized by increased permeability. Within the inter-fault zones, relatively weakly disturbed areas have been found, from which two sites were selected for their promise as possible repository locations. After detailed studies, a site for construction of the underground research laboratory is to be chosen with the prospect of its subsequent conversion to the underground repository. Krasnoyarsk Region A team of experts representing Ministry of Atomic Energy institutions, the Russian Academy of Sciences, and other organizations has examined where to locate a HLW repository in the region of the Krasnoyarsk MCC. Such a repository would be de- 10 Volcanic rocks at the PA “Mayak” region were subjected to metamorphic overprint that resulted in reduction of their permeability. The average sample permeability of the host rocks section is ~10–19 m2, porosity is 0.4% (Petrov et al. 1998). The data on transport properties of the Far East region basalt seria are not yet available and will require study.
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signed to accept solidified HLW and SNF from the Krasnoyarsk MCC, including wastes from RT-2 if it is completed and operated. The examination was carried out by using a stepwise approach: from the stage of searching for promising geological formations to the stage of choosing promising sites. Based on the results obtained from research to date, and taking into account socio-economic and environmental factors, the Nizhekansky granitoid pluton was selected as a candidate rock massif. The potential host rocks are biotite granites and granodiorites. The studies of the massif helped to identify several promising sites with low rock permeability and high tectonic stability. Geological and geophysical work provided a basis for selecting the two most promising sites: the “Itatskiy” and the “Kamennnyi” sites, each with an area of 7 km2, and both located about 25 km to the southeast of the Krasnoyarsk MCC. At present, investigations for choosing the most promising site for designing an underground research laboratory are being conducted. Northwest Region An international team of experts, operating under the collaborative Russian-European Tacis project, completed a study in 2001 of issues related to interim storage of SNF and disposal of radioactive waste from operating and decommissioned nuclear-powered submarines in the northwest region of Russia, as well as from the Kola Nuclear Power Plant. The study included selection of storage and disposal sites, facilities arrangement, and technical equipment. Expert participants in implementation of this project were from the Russian Academy of Sciences, Minatom RF, scientific-research institutions of the Arkhangelsk oblast, as well as experts from Belgium and France. At the first stage of the project implementation, 25 candidate sites were selected within the region. From these, after analysis, 7 sites were recommended for more detailed assessment. After additional studies, at the closing stage of the project, two sites were selected as potentially favorable for SNF long-term, dry, on-surface storage and for geologic disposal of radioactive waste. One site is located near the Navy bases and enterprises for dismantling, salvage, and disposition of nuclear-powered ships at the northern, low-seismicity coastal zone of the Kola Peninsula. The second site is at the southern, low-seismicity coastal zone of the Kola Peninsula. Both sites are formed by old
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crystalline rocks dominated by granitoids, granite-gneisses and migmatites characterized by high strength, low permeability, and insignificant tectonic disturbance. At present, financing is needed for engineering studies at the selected sites, for development of the on-surface infrastructure, and for preparation of mining works for construction, equipment, and operation of an underground research laboratory with the prospect of its subsequent conversion to an underground repository. Far East Region In the Far East region in the southern part of the Primosky Krai, the Artemovsky site, located at a distance of 70–80 km to the northeast of Vladivostok city, may be recommended for the underground interim dry storage of SNF and other radioactive materials from nuclear-powered submarine (NPS) operation and decommissioning. The territory of the region is formed by terrigenous-sedimentary coal-bearing clayey-sandstone neogenic rocks covered by the mantle of tholeiitic plateau basalts and neck facies of alkaline basalts. The total thickness of volcanic formations reaches 300 meters, and their absolute age is about 4 million years. Plateau basalts and alkaline basalts have appropriate physical and geochemical isolation properties and may be utilized as a host media for construction of a facility for SNF underground dry storage. The sharply rugged topography of the Artemovsky site with the altitudes varying from 600–700 m up to 1,200–1,250 m suggests that horizontal drifts, which can be used as access and emplacement tunnels, would be the most reasonable construction option for a storage facility. Such construction has significant technical and economical advantages in comparison with a shaftfed facility, as it eliminates the need for the shaft well with the lowering and lifting equipment, and simplifies and reduces the cost of the water pumping and ventilation. The relative proximity of the Artemovsky site to the city of Vladivostok—a large administrative and industrial center—and the presence of a developed transportation network in the territory make it a potentially desirable host area for construction of the underground storage facility.
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Transbaikal Region The Priargunsky Industrial Mining Chemical Association (PIMCA), located in the southeastern Transbaikal egion (Chitinskaya oblast, town of Krasnokamensk), is under Minatom administration. PIMCA is the only enterprise in Russia carrying out mining and processing of uranium ores, and it is in a region that is a promising site for an underground interim dry storage facility for SNF and other radioactive materials. The region is located at a significant distance from large settlements and industrial centers, but is connected to other regions of Russia by a railway line and automobile roads. The region under consideration is a low-seismicity zone. Most of its territory is formed by massive crystalline rocks with high mechanical stability, low permeability, effective isolation properties, and weak tectonic disturbance. Like the Artemovsky site, the rugged topography, with altitudes varying from 600 m up to 900–1,000 m, allows horizontal access and emplacement tunnels. Collection of additional materials at the selected sites is needed for a clearer understanding of the geological, hydrogeological, geophysical, and other conditions for development of underground SNF and HLW storage and HLW disposal facilities (Velichkin et al 2002). 3.3.2 Nuclear Waste Underground Disposition and Disposal in the United States At the Hanford site, the SRS, and the West Valley Demonstration Project, the acidic liquid effluent was neutralized with sodium hydroxide for storage in carbon-steel tanks. This neutralization process produced a metal-rich precipitate known as sludge. To conserve tank space, the HLW at Hanford and Savannah River was concentrated using evaporators to drive off excess liquids. This produced a salt-rich slurry that if sufficiently concentrated, crystallized into a solid salt cake upon cooling. As a result, the HLW in storage at Hanford and Savannah River exists in several physical forms: liquid, salt cake usually containing interstitial liquid, slurries (liquids with suspended particles), and sludge. Over 98 percent of the roughly 3.5×105 m3 of tank wastes are aqueous liquids or slurries (DOE 2001b). Current efforts to immobilize HLW at SRS and Hanford for disposal are greatly complicated by the waste’s physical and chemical heterogeneity.
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The HLW stored in the tanks at Hanford is especially complex as a consequence of its history of production and management. Prior to about 1952, reprocessing at Hanford was carried out using a bismuth phosphate process, which produced dilute wastes that contained high concentrations of uranium. After the REDOX and PUREX processes were introduced starting in the early 1950s, the HLW was more concentrated and, in some cases, became self-boiling after being pumped into the underground tanks (in fact, heat loads were sufficient to damage some of the tanks). To reduce heat loads, cesium and strontium were removed from the HLW by chemical precipitation and ion exchange processes. The separated strontium and cesium were loaded, in the form of halide salts, into steel capsules used as irradiation sources onsite and offsite. These high-intensity sources are currently stored at Hanford. These materials will presumably be disposed in a geologic repository, but the exact disposition pathway is unclear at present. Experience in the United States has shown that storing liquid HLW in underground tanks for decades past their design life is unreliable and hazardous. Physical and chemical processes in the waste result in waste forms that are difficult to manage. Leaks from degraded storage tanks have resulted in plumes of contamination that are hazardous and difficult to clean up. The liquid HLW at INEEL was handled and processed differently from the waste at Hanford and SRS. After production, HLW was temporarily stored in stainless steel tanks and then processed in a fluidized-bed chemical reactor to produce a granular ceramic, referred to as calcine. This calcine HLW is stored in stainless steel bin sets within concrete vaults that are designed to last for 500 years. HLW is to be disposed of by DOE in a deep geologic repository after it has been put in a form suitable for disposal. Immobilization in borosilicate glass (vitrification) is the waste form that has already been selected for HLW at SRS and the West Valley Demonstration Project. Each of these sites has its own vitrification facility. The Defense Waste Processing Facility (DWPF) at SRS, which uses a joule-heated melter, is the largest HLW-vitrification facility in the world. The facility began vitrifying radioactive waste in 1996 and expects to vitrify all of the HLW currently stored at the SRS in 20 to 25 years, producing the molten-glass waste form into approximately 6,000 stainless steel canisters. As of May 2002,
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almost 1,300 canisters of HLW glass have been made and are stored in an underground storage vault in the Glass Waste Storage Building (WSRC 2001). The West Valley Demonstration Project started to vitrify its liquid high-level radioactive wastes in 1996 and completed its efforts in 2002. The West Valley facility used a joule-heating melter to produce borosilicate glass to immobilize the waste from that reprocessing plant. As of August 2002, 275 canisters of vitrified HLW had been made and stored in racks in the High-level Waste Interim Storage Facility. Vitrification is being examined for HLW at the two other major sites: the Hanford Site and INEEL. Other technologies for immobilization are also under consideration. A program has been struggling for several years to develop a vitrification facility at Hanford, called the Waste Processing and Immobilization Facility, to retrieve and immobilize some of the high-level wastes in the 28 double-shell tanks. This will include most of the liquid from the 149 single-shell tanks, which is being pumped into the double-shell tanks. Plans for the solid and semi-solid wastes remaining in the single-shell tanks are still being developed. Construction of the vitrification facility began in 2002, and vitrification of radioactive material is to begin in 2007. Immobilizing the waste is expected to take about 30 years. Waste will be stored onsite prior to shipment for disposal at a mined geologic repository. Some residual contamination will remain in the tanks and substantial quantities of low-activity waste will be generated in the pretreatment and immobilization process. DOE does not consider these residual and low-activity wastes to be HLW and is seeking alternatives for managing these wastes. Plans are still being developed for the calcine HLW and salt-bearing wastes at Idaho. The calcine HLW will be converted to another form and, following treatment, it is expected to be sent to a geologic repository for disposal.
Representative terms from entire chapter: